Microstructural Study on the Stress Corrosion Cracking of Alloy 690 in Simulated Pressurized Water Reactor Primary Environment
This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360 °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.
KeywordsAlloy 690 SCC PWR Intergranular oxidation Boundary migration
The authors gratefully acknowledge financial support through the DOE I-NERI program contract 2011-01-K. The authors would like to thank Young Suk Kim and Sung Soo Kim from Korea Atomic Energy Research Institute for providing the materials for this study, and Alex Flick from the University of Michigan for his assistance with preparation of the high temperature autoclave systems. This research was performed, in part, using instrumentation provided by the Department of Energy, Office of Nuclear Energy, Fuel Cycle R&D Program and the Nuclear Science User Facilities.
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